Concentration and decontamination of solutions containing plutonium values by bismuth phosphate carrier precipitation methods



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Patented as, iaso 25 Claims. (Cl. 23-145) This invention relates to the separation and concentration of transuranic elements contained in dilute solutions thereof or in mixtures with larger amounts of other metal compounds. The invention is especially concerned with the concentration of compounds of element 93 and element 94 contained in less than millimolar concentrations in solutions derived from neutron-irradiated uranium and with the separation of contaminating radioactive elements of lower atomic weights from such solutions of elements 93 and 94.

An object of the present invention is to provide a process for separating lower atomic weight elements from elements 93 and 94 and for concentrating elements 93 and 94 in solutions so dilute as to make impractical the direct recovery as a precipitate consisting solely of insoluble compounds of elements 93 and 94.

Another object of this invention is to provide a multistage process for decoutaminating and concentrating .element 94 in such solutions to an extent suflicient to enable the recovery of a final precipitate of a substantially pure compound of element 94. a

An additional object is to provide a method for recovering radioactive fission products from a mixture of neutron-irradiated uranium, transuranic elements and radioactive fission products.

A further object of the present invention is to provide a method for decontaminating and concentrating element 94 present in mixtures containing ions of element 94 and cations of coprecipitable carrier compounds, whereby the ratio of carrier cation to element 94 may be substantially reduced.

Additional objects and advantages of this invention will be evident from the following description.

The term element 94 is used throughout this specification to designate the element having atomic number 94. The designation 94 refers to the isotope of element 94 having a mass number of 239. Element 94 is also referred to in this specification as plutonium, symbol Pu. Likewise, element 93 means the element having atomic number 93, which is also referred to as neptunium, symbol Np. Reference herein to any of the elements is to be understood as denoting the element generically, whether in its free state or in the form of a compound, unless otherwise indicated by the context.

The apparent discovery of transuranic elements (element 93 and elements of higher number) was first announced by E. Fermi in 1934. At that time Fermi stated that the bombardment of uranium with neutrons gave beta activities which he attributed to transuranic elements of atomic number 93 and possibly higher. From 1934 to 1938 other workers, notably Hahn and Curie, extended this work. But in 1939 Hahn discovered that the elements which he and others had believed to be transuranic elements were in fact radioactive elements of medium atomic weight produced by the fission of uranium. Hahus results were subsequently confirmed,

and a great many fission products in addition to those first found by Hahn were discovered and identified.

Such products were all of lower atomic number than uranium, generally of atomic numbers in the middle of the periodic group; and so far as is known, prior to about June 1940 no evidence was found to establish the ex istence of any transuranic element.

However, in June 1940, McMillan and Abelson published in The Physical Review, 57, 1185, their discovery that-a 2.3-day activity produced by the bombardment of uranium with neutrons was an isotrope of element 93, probably 93 Although McMillan and Abelson surmised that element 94 would be formed by beta decay of element 93 they were unable to produce any evidence of its existence and did not obtain either 93 or 94 in pure form or in microscopic amounts, either as the element or as a compound.

A. C. Wahl, G. T. Seaborg and l. W. Kennedy, using the methods of McMillan and Abelson, obtained 93 admixed with rare earths, proved that 93 decayed to 94 and measured the radioactive and fission properties of 94 Subsequently, neutronic reactors were developed for the production of 93 and 94 in insols quantities by a self-sustaining chain reaction of slow eutrons with U and U in natural uranium.

Natural uranium comprises largely isotope U together with about /139 as much U and a very much smaller abount of U When this mixture of isotopes, either as metallic uranium or as a uranium compound, is subjected to bombardment by neutrons from an external source, or undergoes a self-sustaining neutronic chain reaction, a number of nuclear reactions take place. Isotope U captures a neutron to form U which undergoes beta decay to form transuranic elements, as

indicated by the following equations:

Substantially all of the fission fragments have mass numbers within the range 77-158, although small quantitles of isotopes of lower and higher mass numbers may result from unbalanced binary fissions, ternary fissions, or other reactions of infrequent occurrence. A very large majority of the fission fragments comprise a light group of mass .numbers 84-106 and a heavy group of mass numbers 128-150.

The various decay products of the initial fission fragments and the initial fragments are referred to herein as fission products. These fission products fall within a range of atomic numbers from about 32 to about 64. The fission products from the light group of fragments referred to above have atomic numbers ranging from about 32 to about 46; and the fission products from the heavy group of fragments have atomic numbers ranging from about 51 to about 64.

The various radioactive fission products have half-lives ranging from a fraction of a second to thousands of years. Those having very short half-lives may be eliminated by aging the material for a reasonable period before handling. Those with very long half-lives do not have sufficiently intense radiation to endanger personnel protected by moderate shielding. On the other hand, the fission products having half-lives ranging from a few days to a few years have dangerously intense radiations which cannot be eliminated by aging for practical storage periods.

, These products are chiefly radioactive isotopes of Sr, Y,

Zr, Nb, and Ru of the light group and Te, 1, Cs, Ba,

will decrease as the ratio of U content to 94 content of themass decreases. For this reason, a neutronic reaction for 94 production is suitably terminated when only a fraction of the U has been converted to fission products. The reaction mass at this point contains a large amount of U a much smaller amount of U still smaller amounts of 93 94 and fission products, and traces of other products such as UX and UX By aging such a mass for a suitable period of time, the 93 may be substantially completely converted to 94 with simultaneous conversion of the short-lived radioactive fission products to longer-lived or stable isotopes.

The plutonium content of the aged reaction mass will usually be less than 1% of the weight of the unreacted uranium and may even' be less than one part per million parts of uranium. The concentration of fission products will be of the same order'of magnitude. The separation of plutonium from such a mass involves extraction from the unreacted uranium, decontamination by separating the radioactive fission products, and concentration of the decontaminated material to obtain a product from which a relatively pure plutonium compound can be directly recovered.

Techniques have been developed for extracting plutonium from asolution of neutron-bombarded uranium by means of a carrier precipitate such as bismuth phosphate or lanthanum fluoride. The precipitates thus obtained are substantially free from. uranium but contain sufficient amounts of fission products to have intense radiations. Similar techniques have also been developed for the decontamination of the extracted plutonium by means of a carrier precipitate. Such methods usually involve alternately carrying plutonium in one state of oxidation with the precipitate, and leaving plutonium in solution in a second state of oxidation While carrying fission products with the precipitate. After decontamination by such methods, the plutonium may be substantially free from dangerously radioactive fission products but is still present in relatively minute amounts associated with large quantities of a carrier precipitate or with large volumes of solution containing ions of the preceding carrier precipitate- V 7 Solutions of plutonium obtained by such procedures have such low plutonium concentrations that the direct precipitation of even the most insoluble plutonium compounds is either impossible or wholly impractical. The application of common concentration procedures to such solutions results in concentration of carrier ions as well as plutonium ions, and increases the difficulties of subsequent plutonium separation.

In accordance with the present invention, the plutonium in dilute solutions containing fission products may be both decontaminated to a high degree and if desired concentrated at the same time. This decontamination and concentration may be brought about by the use of alternate carrier precipitatesof different chemical compositions in which one carrier comprises a precipitate of bismuth phosphate. In this manner, adequate decontamination may be secured by fission-product-carrying precipitations between plutonium-carrying precipitations, and the ratio of carrier to plutonium may be successively decreased in each plutonium-containing precipitate. Each plutonium-carrying precipitate may be dissolved in a smaller volume of solution than the preceding precipitate; and a solution may finally be obtained from which a plutonium compound may be precipitated without any carrier. The term carrier" as used herein and in the appended claims signifies a substantially insoluble, solid, finely divided compound having an inorganic cation and to yield at least one anion which forms a compound with the ion to be carried which is at least as insoluble as said carrier.

A large number of carriers are available for use in the process of the present invention. Numerous individual carriers have previously been used in extraction and decontamination processes, and combinations of these carriers are suitable for use in the present inventive process. The following are representative examples of useful carriers:

It will be apparent that the above compounds constitute plutonium carriers in accordance with the definition previously given. Thus, bismuth phosphate is capable of ionizing to form a bismuth cation and a phosphate anion. The latter is an ionic component of the insoluble compound Pu '(PO In like manner, the compound sodium uranyl actate is capable of ionizing to yield the cations Na+ and (UO' and the anions [UO (C H O l" and (C H O 'The acetate anion is common to the insoluble compound sodium plutonyl acetate. In a' similar manner, the compounds listed above may serve as fission product carriers. Thus, the phosphate anion of bisrnuth phosphate can serve as an ionic component of insoluble compounds of fission products, e.g., strontium, yttrium, and zirconium phosphates, while simultaneously serving as an ionic component of a soluble compound of hexavalent plutonium, i.e., plutonyl' phosphate.

In carrying out the present invention, bismuth phosphate is precipitated from a dilute solution of plutonium and fission products; the precipitate is redissolved to form a solutionof reduced volume; at. least one fission product carrier is precipitated and separated from this solution; and then a second plutonium carrier, difiering in chemical composition from bismuth phosphate is precipitated and separated from the solution. After one or more cycles of this procedure, a plutonium-carrying precipitate may be obtained which is substantially free from fission products and which may be dissolved in a 'sufiiciently small volume of solution so that a substantially pure plutonium compound may then be precipitated free from carrier. V i

In any cycle of the present process, the intervening fission product'carriers may be the same as one or both of the plutonium carriers, or .they may be chemically 'distinct from both of the plutonium carriers. From the standpoint of decontamination, it is generally desirable to follow a plutonium carrier with the same carrier, or one isomorphous therewith, as a fission product carrier. Conversely, more complete decontamination may be secycle if the plutonium carrier'is isomorphous with the plutonium compound tobe carried but not isomorphous with thepreceding fission product carrier. However, if two intervening fission product carriers are employed, it is generally more convenient to use the same carriers as the preceding and subsequent plutonium carriers.

In order to employ the same compound successively as a plutonium ca'rrier and as. afission product carrier, conoured in the final plutonium-carrying precipitation of the.

fitions must be employed which prevent the carrying of plutonium in the fission-product-carrying precipitation. This may conveniently be accomplished by changing the state of oxidation of the plutonium. Thus, the plutonium may be maintained in solution in a higher state of oxidation while precipitating a fission product carrier in a lower Valence state. Conversely, the plutonium may be maintained in solution in a lower state of oxidation while precipitating a fission product carrier in a higher state of oxidation.

Plutonium has a number of valence states, including +3, +4, +5, and +6. In 0.5-4.0 M aqueous hydrochloric acid, the oxidation-reduction potentials are of the following magnitudes:

. Volts Pu+ +Pu+ +E -0.97 Pu+' Pu+ +E- 1.11- Pu+ +2H- O- PuO +4H++E- -0.92 .}Pu+ +2H O PuO +4H++2E-' 1.015

The Pu+ ion is generally very unstable and, in view of experimental evidence at present avialable, is believed to disproportionate to Pu+ and Pu Also, from additional experimental evidence, the Pu+ ion is believed to be capable of disproportionating to the Pu+ ion and the PuO ion, and in aqueous hydrochloric acid this disproportionation may take place to a considerable extent. Such disproportionation of the Po is opposed, however, by increase in hydrogen ion concentration and by the presence of ions which tend to complex or other-' wise stabilize the Pu+ ion. The effect of additional ions in hydrochloric acid solutions is illustrated by the following potentials for the Pu+ Pu+ couple:

Volts 1.0 M HCl --0.97 1.0 M HCl-O.1 M H3PO4 0.80 1.0 M HCl-1.0 M HF -l +0.53

Generally the anions of slightly ionized acids tend to complex the Pu+ ion to a much greater extent than the anions of highly ionized acids. Thus, Pu+ is only slightly complexed by C10,, Cl-, and N0 it is complexed to a greater extent by S05 and it is very strongly complexed PO4 3, F C2H302", and (2204 In addition to the complexing effect of the anions of the acids employed as solvents for plutonium, certain of these acids may also serve as oxidizing agents. However, at room temperatures, or moderately elevated tempera tures, and in the absence of oxidation catalysts, the rate of oxidation by the acid is often so low that this etfect may be ignored. Thus, the Pu+ ion is stable for considerable periods of time in perchloric acid, although under proper conditions, the latter is capable of oxidizing Pu+ to PuO+ It is therefore desirable to control the state of oxidation of the plutonium by the use of oxidizing and reducing agents which have rapid reaction rates under the conditions employed for processing the solutions.

The. Pu+ ion may suitably be oxidized to the PuO ion by the addition of an active oxidizing agent having.

an oxidation potential substantially more negative than the oxidation potential of the PuO Pu+ couple in the particular solution employed. The following are representative potentials for this couple:

Volts 1.0 M HCl 1.0 1.0 M HNO 1.1 1.0 M H 80 1.3

book of Chemistry and Physics (Chemical Rubber Publishing Co., Cleveland, Ohio).

It is generally preferred to employ solutions of 'plutonium in aqueous nitric acid. As examples of oxidizing agents for use in such solutionsthere may be mentioned the alkali metal bismuthates, the alkali metal dichromate's, and lead dioxide, either as PbO per se, or as Pb O To efiect the oxidation, a quantity of oxidizing agent at least equivalent to the amount of plutonium is added to the solution, and the resulting mixture is digested at a moderately elevated temperature for a sufiicient period of time to insure complete oxidation of the plutonium. In most cases, this digestion may suitably be effected at 6080 C. for 15-60 minutes. In order to maintain the plutonium in the hexavalent state for considerable periods of time after oxidation, it is desirable to employ an excess of oxidizing agent to serve as a holding oxidant. This is especially true if an acid solution is to be processed in ferrous metal equipment, or under other conditions permitting subsequent reduction of the plutonium.

For thereduction of plutonium, reducing agents of adequate potential may be selected by reference to tables of standard potentials such as the table previously referred to. For the reduction of Pu0 or Pu+ to Pu+ the reducing agent should have a reduction potential substantially more positive than the reduction potential of the Pu+ Pu+ couple in the solution employed. hus, in 1.0 M HCl an active reducing agent having a potential more positive than 0.97 v. Will be required, and in 1.0 M HNO a potential more positive than 0.92 v. will be necessary. In order to maintain the plutonium in the +3 valence state for appreciable periods of time, it is desirable to maintain an excess of the reducing agent in solution.

In order to reduce Pu0 to Pu+ Without reducing Pu+ to Pu+ it is desirable to employ an active reducing agent having a reduction potential substantially more positive than the reduction potential of the Pu+ -+PuO couple, and substantially more negative than the reduction potential of the Pu+ Pu+ couple, in the solution used. A wider selection of reducing agents of the desired potential will be available for use in solutions containing ions which'cornplex the Pu+ ion than is available for use in solutions which are substantially free from complexing effects. Thus, in 1.0 M HCl and 1.0 M HCl1.0 M HF, the reduction potentials are approximately:

1.0 M H01 1.0 M HG1 1.0 M HF Pu+ 1 uO2+ 1.0 v -1.2 v. Pu+ Pu+ 0.9 v. -O.5 v.

. state in either solution. When employing the preferred solutions of plutonium in aqueous nitric acid, the reduction of PuO to Pu is preferably eifected in the presence of a complexing ion, employing reducing agents having reduction potentials of the same magnitude as hydro v gen peroxide and ferrous iron. However, it is also possible to use stronger reducing agents such as sulfur dioxide if any excess reducing agent is removed or destroyed after the initial reduction is effected. In any case, if Pu is desired, the hydrogen ion concentration should be sufiiciently high to oppose the disproportionation of Pu+ to Pu+ and PuO For this purpose, it is desirable to employ solutions having a pH not substantially above 2, and preferably considerably below 2. In the case of aqueous nitric acid solutions it is generally desirable to maintain a free acid concentration of at least 1 M.

It will be apparent that the considerations discussed above will also apply to the oxidation of Pu+ to- Pu without oxidizing Pu+ to PuO by the use of oxidizing agents having potentials intermediate the potentials of the twoplutoniu'm couples.

Since the anions of most of thepreferred carriers are capable of complexing the Pu+ ion, the complexingof the plutonium may be eifected by the carrier anion when carrying plutonium in the +4 valence state. The phosphate, fluoride, and oxalate carriers are particularly advantageous in this respect. In order to provide adequate complexing, it is desirable to employ an excess of the In the preferred modification of our process, two or more fission product carriers of difierent chemical composition are employed between the initial bismuth phosphate plutonium carrier and the final plutonium carrier. In this manner maximum decontamination as Well as maximum concentration may be accomplished in a single cycle of the process. Suitable carrier combinations for two fission product carrier precipitations between the first bismuth phosphate carrier step and the second pluto- 1'0 m'um carrier precipitation of a cycle are set forth in Table carrier anion. This is also advantageous in improving H:

I TABLE 11 Initial First j Subsequent First Fis- Plutonium Plutonium Metathesizlng Solution of oxidizing Agent sion Prod- Solution Carrier Agent Plutonium not Carrier Precipitate Preeipitate Aq.HNO3 BiPO AqJINOa--- NaBiOa-i-K;Om01 BiPO4. AQ.HNO3 BiPO AILHNOz... PbgO BlPO Aq.HNO3 BiPO NaOH+K OO3 Aq.HNO9 NaBiO KzOIzO BiPO Aq.HNO3 BiPO Aq.HNO3 1 11304." BiPO Aq.HNO3 BiPO NaOH A HNOQ-.. Ce(Noa)4 CePO4. BiPO4.

Second Fission Reduc- Second Plutonium Final Solu- Product Carrier ing Carrier Precipitate Methatheslzing Agent tion of Precipitate Agent Plutonium LaFa Na0H+KnCOi AqHNOa Th(IOa)4 Aq.HC1 Nb O; AqQHzSO; NaUOz(C:HsO2)a Aq.HNOa LaFg ACLHNO:

the degree of carrying when the carrier is not isomorphous with the plutonium compound to be carried. The excess carrier anion may suitably be present in concentrations ranging from 0.01 M to 1.0 M.

In the preferred process of the present invention, the plutonium is carried in the +4 valence state and is maintained in solution in the +6 valence state while carrying fission products. In accordance with one modification of this process, an operating cycle comprises the precipitation of bismuth phosphate thereby carrying down the +4 plutonium present, solution of the precipitate, oxida tion of the plutonium to the +6 state, precipitation of a fission product carnenwhich may. be bismuth phosphate or another suitable carrier, reduction of the plutonium to the +4 state, precipitation of. a second-'plutonium carrier chemically distinct from bismuth phosphate, and solution of the second precipitate toform a smaller volume of solution than that resulting from the dissolution of the first precipitate. Suitable carrier combinations for a single fission product carrier precipitationintermediate two plutonium carrier precipitations are set forth in Table I:

59 in g carrier.

Although carriers may be be employed as pre-formed finely divided solids, it is generally preferable to precipitate 35 the carrier in situ since the latter procedure usually permits a lower carrier ratio and results in more nearly quantitative carrying of plutonium or fission products. Substantially the same techniques forcarrier precipitation. may be employed in the present multiple carrier 40 process process as have previously been used in'single carrier extraction or decontamination procedures. In general, it is desirable to incorporate the carrier cation in the solution; agitate while adding the carrier anion, and digest the resulting mixture prior to separating the preci pitate.

When changing carriers in the present process, the preceding precipitate is suitably dissolved in the minimum volume, of solution from whichthe subsequent carrier may be precipitated substantially free from the preced The use of different solvents in succeeding stages will facilitate volume reduction, but the same solvent may be used if the concentrations are suitably ad justed. It is generally preferred to employ aqueous solvents and to modify their solvent power from stage to TABLE I First Plu- Initial Plutotonium Metathe- Subsequent Solution nium Solution Carrier sizing lutonium Oxidizing Agents Prccipi- Agent ate BlPO4... A'ILHNO: .NaBi0a+K2Cr2O1. BiPO Aq.HNO3 NaBiOa+K2CrzO7. BlPO4 Aq.HNO KzSaOH-AgNOa. B1PO4 Aq.HNO3- 0e(N03)-1+K2Cr207. BiPO4 Aq.HNOa.. NaBiO3+Oe(NO;)4. BlPO 1120204. AQ.(NII.|)2C20{ p Fission Product Carrier Reduc- Second Plutonium Metathesizing Final Solu- Precipitate ing Carrier Precipitate Agent tion of Agent Plutonium .NaOH+K C O3-.- Aq.HNO3. NaOH-l-KzCOa--- ACLHNOZ. Aq.HN03.

Aq.HC1. BlPOt ACLHNOz. BIKCQOQQ Aq-HNOa.

stage by adjustment :of ionic concentrations. Thus, an aqueous solution of an inorganic acid or base may be used as the solvent in successive stages of the process and the pH may be adjusted to increase the solvent power from stage to stage. Alternatively, an additional ion may be introduced to form a soluble complex with the cation of the preceding carrier. Other equivalent procedures for reducing the volume of solution from stage to stage and for precipitating a carrier free from preceding carrier will be evident to those skilled in the art.

The ratio of carrier to plutonium in the plutonium carrying steps of the present process may vary over a wide range depending on the plutonium concentration of the original solution and upon the effectiveness of the particular carrier employed. Ratios ranging from 10,000/ 1 or higher in the first stage of the process to /1 or lower in the final stage may be used. However, the ratios '-.\vill generally fall within the range 1,000/1 to 100/1.

The amount of fission product carrier in any step of the process may suitably be of the same magnitude as the preceding plutonium carrier.

In the final plutonium-carrying precipitation, a relatively low ratio of carrier to plutonium is desirable, and in such case an isomorphic carrier is preferred. An isomorphic carrier is one having a crystalline structure with cation spacing in the crystal lattice such that plutonium ions may be substituted in the lattice for carrier cations. Since cations of isomorphic compounds tend to precipitate out of solution together irrespective of the anion used, it is preferred not to employ isomorphic carriers for two successive precipitations in which plutonium is in the same valence state.

It is apparent that different carriers will be required for the isomorphic carrying of plutonium in its different valence states. For plutonium in the +6 valence state, uranyl compounds are suitable isomorphic carriers; for plutonium in the +4 state, uranous, eerie, and thorium compounds are isomorphic carriers; and for plutonium in the +3 state, cerous and lanthanum compounds are isomorphic carriers. It is generally preferred to maintain plutonium in the +4 state in each of the plutoniumcarrying steps of the process. In addition to the particular carriers mentioned above for Pu other isomorphic carriers of the same range of ionic diameters may be employed. Useful carriers comprise those having tetravalent cations of ionic diameters within the range 2.17-2.36 A., as corrected in accordance with Zachariasens method for determining corrected ionic radii (Zeit. fiir Kryst. 80, 137,- 1932).

After one or more cycles of plutonium carrier and fission product carrier precipitations in accordance with the foregoing procedure, a final precipitation may be made with a sufiieiently low ratio of carrier to plutonium so that the precipitate may be dissolved in a small volume of solution and a plutonium compound may then be precipitated directly, without a carrier. If an isomorphic carrier is employed in the final carrier stage of the process, it will be necessary to change the valence state of the plutonium, or of the carrier cation, in the final solution in order to make a final precipitation of a plutonium compound free from carrier. On the other hand, if the final carrier is non-isomorphic, it will only be necessary to select conditions for the final precipitation of the plutonium compound such that at least the cation of the carrier remains in solution.

In a preferred embodiment of this invention the extraction and decontamination of plutonium are carried out in accordance with the foregoing procedure, with a final crossover from a bismuth phosphate carrier to a lanthanum fluoride carrier. By the cross-over step, the ratio of plutonium to carrier, may be greatly increased with ratio changes of the order of 1-10,000 for plutonium-bismuth phosphate to 1-100 for plutonium-lanthanum fluoride. The combined plutonium and lanthanum fluorides are then converted to the hydroxides by metathesis with a suitable base, such as sodium hydroxide, and a final separation of the plutonium is obtained by converting the hydroxides to the peroxides in an acid solution, whereupon the plutonium peroxide will precipitate.

Our invention will be further illustrated by the following specific example:

Example I A dilute solution of plutonium, obtained by separating unreacted uranium and a portion of the radioactive fission products from a solution of a neutron-bombarded uranium, is concentrated and further decontaminated by the following procedure:

The plutonium solution, comprising 3,123 gals. of 0.9 N aqueous nitric acid containing about 0.0003 lb. of plu tonium per gal. as plutonous ion (Pu+ is introduced into the first plutonium precipitator. To this solution is added 39 gals. of a bismuth nitrate solution comprising 1.25 lb. per gal. of bismuth (Bi+ in 10 N nitric acid, while agitating and maintaining a temperature of about 75 C. Approximately 138 gals. of 75% orthophosphoric acid is then added, with continued agitation, after which the mixture is digested at 75 C. for one hour. The bis muth phosphate-plutonous phosphate precipitate is then separated by centrifuging and is washed with 20 gals. of an aqueous solution containing 6.3% nitric acid and 2.9% orthophosphoric acid. The precipitate is next slurried with 275 gals. of 48% nitric acid, and the slurry is washed from the centrifuge to the dissolver and oxidation reactor with 20 gals. of water. The mixture is then agitated to dissolve the precipitate, and the resulting solution has a volume of about 295 gals, and is approximately 9 N with respect to nitric acid.

To the above solution there is added 231 gals. of water and 13 gals. of a- 10% aqueous solution of sodium bismuthate. The solution is then digested at 50 C. for one hour, with agitation, to complete the oxidation of Pu+ to PuO Approximately 7.7 gals. of 10% aqueous K Cr O is introduced as a holding oxidant and the solution is then transferred to the first fission product precipitator where it is diluted with about 2148 gals. of water and heated to 75 C. About 30.4 gals. of 75% aqueous H PO is then introduced while agitating the solution. The resulting slurry is agitated for one hour at 75 C. and is then cooled to 35 C. and centrifuged to separate the bismuth phosphate-fission product precipitate.

The supernatant liquor from the centrifuge, comprising about 2725 gals. of solution in which the plutonium is present as PuO ion, is then transferred to the second fission product precipitator. Approximately 204 gals. of a 1% aqueous solution of LaNH (NO .10H O is introduced and 63 gals. of anhydrous HP is then added with agitation. The resulting slurry is digested for one-half hour at a temperature below 35 C. and is then centrifuged to separate the lanthanum fluoride-fission product precipitate. The supernatant liquor from the centrifuge, comprising about 2992 gals. of solution which still contains the plutonium as PuO ion, is then transferred to the reduction reactor. About gals. of 27.5% H 0 solution is added and the mixture is agitated for one hour at a temperature below 35 C. to complete the re duction of PuO to Pu+ The reduced plutonium solution is transferred to the second plutonium precipitator where it is maintained at room temperature and agitated while adding 58.2 gals. of a lanthanum solution and 14.8 gals. of anhydrous hydrofluoric acid. The lanthanum solution was prepared by forming a solution of 358 lbs. water and 83 lbs. of 60% nitric acid, followed by the addition of 56 lbs. of La(NO .2NH NO .4H O with agitation. The resulting slurry is digested at a temperature not substantially above 35 C. for one-half hour, and the lanthanum fluoride-plutonous fluoride precipitate is then separated by centrifuging and is washed with water.

The fluoride precipitate is next treated with 6 gals. of 40%: sodium hydroxide solution, and the slurry is transferred to'the metathesis reactor and dissolver where it is agitatedfor one-half hour at 5060 C. An additional 6 gals. of 40% sodium hydroxide solution is then added and the agitation is/continued for another half-hour at 50-60 C. The supernatant solution'is then separated and the La(OH) -Pu(OI-I) mixture is thoroughly washed with water to remove'residual sodium hydroxide and sodium fluoride. The hydrated hydroxides are then dissolved in about 2 gals. of 30% nitric acid to produce a solution having a volume of about 8 gals.

It may be seen that the above process effects a concentration ofthe plutonium solution to approximately 0.26% of the volume of the initial solution. The final ratio of carrier to plutonium is about 0.10% of the initial ratio; and the final fission product concentration is about 0.31% of the'initial concentration, based on total gamma radiation of the initial and final solutions. The final solution has a plutonium concentration sufficiently high to permit direct precipitation of plutonium peroxide, or other insoluble plutonium compound, without thew-precipitation of a carrier.

This final separation may be achieved by adjusting the acidity of the solution to approximately 0.01 N (or a pH value of between 1 and 2), and the addition of hydrogen peroxide to give a percent concentration. Under these conditions the plutonium precipitates as the peroxide. The separation of the plutonium by the process of this-invention gives a recovery of greater than 99 percent;

Another embodiment of this phase of this invention is concerned with obtaining a final plutonium carrier which may be easily dissolved in an acid solution. in order to accomplish this object it is necessary that the product carrier fulfill the general requirement of a crossover carrier. It must carry plutonium ions efiiciently and the precipitate must dissolve in a smaller quantity of solvent than the preceding solution so that the plutonium will be present in sufficient concentration that it may be precipitated directly from the solution. We have discovered that lanthanum phosphate issuch a carrier. Experiments have shown that a lanthanum phosphate precipitate will carry the Pu+ ion with greater than 98%. ethciency for a single. cycle. A lanthanum phosphate carrier may be dissolved in an acid solution of much less volume than it is precipitated from and plutonium may be concentrated efiectively by a cross-over precipitate of lanthanum phosphate. Lanthanum phosphate while insolublein a basic solution or one less acid than approximately 0.3 N HNO is readily soluble in solution of higher acidity. Determinations of the solubility of lanthanum phosphate in nitricacid'solutions at 25 C. have shown that at 1 N nitric acid concentration the solubility is approximately 18 gm./l. The solubility increases to approximately 185 gm./l. in 5 N nitric acid solution and to approximately 500 gm./l. in a N solution.

In accordance with the embodiment of our invention employing lanthanum phosphate as a cross-over carrier in the plutonium concentration step, the plutonium is carried through the extraction and decontamination steps,

by means of oxidation-reduction procedures, with hismuth phosphate and other carriers, substantially as described above. Following the decontamination steps plutonium is normally present in a nitric acid solution in the +4 oxidation state together with Bi+ and PO ions. Before precipitating lanthanum phosphate in this solution it is necessary that the excess bismuth ions be removed. This may be accomplished by oxidizing the plutonium to the +6 state, and separating a bismuth phosphate precipitate, then reducing the plutonium to the +4 state. A lanthanum phosphate carrier is then precipitated in a solution by contacting the solution with a. soluble lanthanum compound'and reducing the acidity 12 of the solution. This precipitate is then separated by any normal means and may be redissolved in an acid solution.

The lanthanum ions may be added to the solution containing plutonium in the +4 state as the chloride or preferably the nitrate. The solution will normally contain an excess of phosphate ion but if it is desired phosphoric acid may be added to the solution. Although the plutonium may be contained in any inorganic acid solution, it is preferable that a nitric acid solution be used since the lanthanum nitrate is more soluble than other lanthanum compounds. NaOH, (NI-I HPO NaHCO and NaC H O have been found to be suitable neutralizing agents. Lanthanum phosphate will be precipitated when the acidity is lowered to approximately03 N HNO but it is usually desirable to add an excess neutralizing agent in order to increase the speed of the action. It has also been found desirable to heat the solution to increase the effectiveness ofthe reaction. The precipitate.

may be separated by any ordinary means such as centrifugation, decantation or filtration.

In another embodiment of this invention shown in Tables I and II, plutonium extraction and decontamination steps are carried out with a bismuth phosphate carrier, with a final cross-over to a sodium uranyl acetate carrier. It has been found desirable to remove the bismuth phosphate from the solution containing the plutonium prior to the formation and precipitation of the sodium uranyl acetate carrier, in order that the formation of a uranyl phosphate precipitate may be avoided. There are two preferable means of accomplishing this. By one method, the acid solution containing the plutonium ion in an oxidized state and the bismuth ions and phosphate ions is contacted with zirconium nitrate, and the acidity is adjusted so that it approximates a pH between 1 and 2, permitting a mixed phosphate of zirconium and bismuth to form, which'may be separated from the solution. It should be noted that if cerium ions are present in the solution as the result of the use of a chromic-ceric oxidizing agent in prior oxidation steps a cerium phosphate precipitate of a thick jelly-like consistency may form which will interfere somewhat with a clean separation. This may be avoided, however, by the use of the chromate ion'alone as an oxidizing agent, 'or preferably by the use of a sodium bismuthate oxidant. Following the separation of the bismuth phosphate, the filtrate containing the plutonium ion in an oxidized state is contacted with suitable reagents so that a carrier precipitate of sodium uranyl acetate is formed. It has been found that a preferable manner of accomplishing this result is the addition of sufiicient uranyl nitrate hexahydrate to make the solution approximately 0.05 M, followed by the addition of sodium acetate, acetic acid, and sodium nitrate. The sodium uranyl acetate precipitate, which has carried the plutonium in oxidized form, is then separated from the solution and dissolved in an inorganic acid such as nitric acid, and the plutonium is reduced by a suitable reducing agent such as hydroxylamine. The cross-over from a bismuth phosphate carrier to a sodium uranyl acetate carrier greatly reduces the ratio of plutonium ion to carrier and a final separation may be effected by the direct precipitation of plutonium from'solution by the formation of an insoluble plutonous salt such as the fluoride, phosphate or peroxide. it has been found that somewhat better yields may be obtained, however, by the formation of a second sodium uranyl acetate precipitate in the solution containing the plutonium ion in a reduced state; In this embodiment, the sodium uranyl acetate precipitate containing plutonium in the oxidized state is dissolved in an mor ganic acid, such as nitric acid, and the plutonium is then reduced by a suitable reducing agent, such as hydroxy; amine. Since plutonium ion in a reduced state is not carried by sodium uranyl acetate, the second sodium uranyl acetate precipitate may be separated from the solution, leaving plutonium ion in solution, substantially free from all contaminating cations. The plutonium may then be precipitated in a substantially pure state, by the formation of the fluoride, peroxide, phosphate, oxalate, or other insoluble plutonous compound, or it may be oxidized and precipitated with an anion forming an insoluble compound with the plutonyl ion, such as the hydroxide.

In an alternate method of separation with sodium uranyl acetate, the extraction and decontamination steps using bismuth phosphate as a carrier precipitate are carn'ed out as previously described, but the bismuth phosphate is separated from the plutonium by means of an adsorption process, described in pending patent applications, U.S. Serial No. 565,989 of John W. Gofman, Robert E. Connick, and Arthur C. Wahl granted as U.S. Patent No. 2,871,251 on January 27, 1959, and U.S.

-- Serial No. 565,988, of Robert B. Dufiield on which US.

Patent No. 2,832,793 was granted on April 29, 1958. By this method, the acid solution containing the plutonium ion in a reduced state and the bismuth and phosphate ions, preferably with the phosphate ions having a concentration of about 0.4 M, is passed through an adsorption column whereby the plutonium ions in the reduced state are adsorbed. Adsorption materials which have been found suitable, include the zeolites, the ortho form of zirconium phosphate, zirconium phospho-silicate, and titanium phospho-silicate. The plutonium may be removed from the adsorption materials by elution with a suitable acid such as 7 M HNO The plutonium ions are then oxidized and a plutonium carrier precipitate of sodium uranyl acetate is formed in solution and separated therefrom. This step may be suitably accomplished by the partial neutralization of the solution with solid NaOH to approximately 1 N, oxidation of the plutonium with a suitable oxidation agent, such as 0.1 M K Cr O or sodium bismuthate, adjusting the acidity 14 Example II Neutron-irradiated uranium was dissolved in nitric acid and the solution obtained was diluted to form a 20% uranyl nitrate hexahydrate solution (also referred to as UNH). This solution contained plutonium corresponding to 278. counts per minute per gram of uranyl nitrate hexahydrate. Following the dissolution, bismuth nitrateand phosphoric acid were added to the solution containing plutonium to form a bismuth phosphate product carrier precipitate which also carried certain fission products. This first bismuth phosphate product carrier precipitate was separated from the uranyl nitrate hexahydrate solution and the precipitate was dissolved in nitric acid. Plutonium which was contained in this solution in the +4 oxidation state was thenoxidized to the +6 oxidation state by making the solution 0.1 M in C1'2O7 and heating to 90 C. Zirconium nitrate was added to the solution and the acidity was reduced to a pH of 1.5 with sodium carbonate, producing a by-product precipitate of bismuth phosphate and zirconium phosphate. This precipitate was separated from the solution. A sodium uranyl acetate product carrier precipitate was then formed in this solution by adding a soluble uranyl salt to make the solution 0.5 M in U0 and adding sodium acetate, acetic acid, and sodium nitrate. The uranyl salt had been treated to free it of UX by contacting the salt with a de-colorized carbon. The sodium uranyl acetate product carrier precipitate was then dissolved in nitric acid. Following its dissolution the plutonium contained in the solution in the +6 state was reduced to the Pu+ state by adding hydroxylamine to the solution. A sodium uranyl acetate by-product precipitate was then formed in the solution in the same manner as described above and separated from the solution, leaving in solution the Pu+ in a comparatively pure state. the various operations and the amount of decontamination are shown in the table which follows.

TABLE HI Alpha Beta Gamma c./m./g. Percent c./m./g. Percent c.lm./g. Percent UN H UN H UN H Original UNH Solution 278 100 1. 32 10 100 1, 926 100 UN H Solution after BiPOi Product Precipifate 0 0 1. 167x10 88 1, 662 86. 5 Solution of BiPOi Product Precipitate 295 100 179x10 13. 5 172 8.9 Product Solution after NaUOzAc Byproduct Precipitate 272 95 640 .048 O 0 to a pH of approximately 1, and adding uranyl ions, sodium acetate, sodium nitrate and acetic acid. The sodium uranyl acetate carrier precipitate formed with its entrained plutonium is dissolved, the plutonium ion is reduced and either separated by a direct precipitation in the plutonous state, or a second precipitate of sodium uranyl acetate is formed and separated, leaving the plutonous ion in solution substantially uncontaminated and in suflicient concentration that it may be separated from the solution by direct precipitation. It alternatively may be oxidized and precipitated in the plutonyl state. De contamination and separation by this method have been found to be very effective, a decontamination factor of 20,000 being achieved.

These embodiments may further be illustrated by the following examples showing the results achieved in experiments using an oxidation-reduction decontamination step with a bismuth phosphate carrier followed by a separation step using a sodium uranyl acetate carrier.

In the following examples the term product carrier precipitate refers to a carrier precipitate containing plutonium, and the term by-product carrier precipitate refers to a precipitate which does not carry plutonium but does carry contaminants of plutonium or other reactants used in the separation process.

in nitric acid to form a 20% uranyl nitrate hexahydrate.

solution. This solution contained plutonium in amount to give 315 counts per minute per gram of uranyl nitrate hexahydrate. A product precipitate of bismuth phosphate was formed in and separated from the solution, as described above, carrying with it plutonium in the +4 valence state, and fission products, such as zirconium and niobium, which form insoluble phosphates. This precipitate was then dissolved in nitric acid and the plutonium contained in the solution was oxidized from the +4 to the +6 valence state as described in Example II. A by-product precipitate of bismuth phosphate was then formed by increasing the phosphate ion concentration in this solution to 0.4 M and the precipitate was removed. The plutonium contained in the supernatant solution Was then reduced from the +6 to the +4 valence state by adding hydroxylamine, and the solution, which also con tained at small amount of bismuth phosphate, was passed through an adsorption column containing zirconium and titanium phosphates. The rate of passage was 30 gallons per square feet per hour. Water was then passed through the column to remove the soluble phosphates, and the The percentage of plutonium carried by precipitate was dissolved in nitric acid, the Pu+ ions were reduced to the Pu+ state with hydroxylamine, and a by-product precipitate of sodium uranyl acetate was formed in and separated from the solution, leaving the plutonium in the solution. The results are shown in the following table.

TABLE IV 7 l5 rier for said plutonium selected from the group consisting of lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate,and niobium oxide, sep-. arating the carrier andassociated plutonium from the s11.- pernatant solution, and dissolving the separated'carrier and plutonium to form a third solution having a volume substantially less than the volume of the initial solution.

2. In a processior the concentration and decontamination of plutonium in a dilute solution containing plutonium in the tetravalent state together with fission product values, the steps which comprise precipitating in said solution a bismuth phosphate precipitate, separating the Alpha M Beta Gamma e./ m./g. Percent c./m./g. Percent e./m./g. Percent UNH UNH UN H Original UNH Solution 315 100 1. 228x10 100 1, 920 100 UN H Solution after BiPOi Product Precipitate 8 2. 54 999x10 81 1, 680 87. 5 Solution of BiPOi'Product Precipita 299 94. 9 116x10" 9. 4 164 8. 5 BiPO4 By-prorlucj: Precipitate 2 06 092x10 7. 5 116 6. 0 Column Feed (Product Solution after BiP O By-product precipitate) 308 97. 8 2, 790 23 (0)? 0 Column Waste Solution 9. 5 2, 550 21 (131% 1.55

Product Eluate 244. 77

It is who understood, of course, that the above examples are merely illustrative and do not limit the scope of this invention, Numerous combinations of carriers other than those of the specific examples may be employed; and two or more diiferent carriers may be used successively in a plurality of concentrating and decontaminating cycles. Various modifications of'the procedures employed in this example will also be apparent to those skilled in'the art.

Although the present invention has been illustrated with particular reference to the concentration and decontamination of plutonium, and is especially useful for this purpose, it should be understood that this invention is equally applicable to the concentration and decontamination of neptunium or of mixtures of neptunium and plutonium. All of the procedures disclosed above may be applied to dilute solutions of neptunium, or to dilute solutions of neptunium and plutonium, as Well as to aged 7 solutions containing only plutonium.

In general, it may be said that the use of any equiya lents or modifications of procedure which would naturally occur to those skilled in the art is included in the scope of the present invention. Only such limitations should be imposed on the scope of this invention as are indicated in. the appended claims. 7 V V This is a continuation-in-part of co-pending application Serial No. 519,714, filed January 26, 1944, on which US. Patent No. 2,785,961 was granted on March 19. 1957., and all subject matter therein not inconsistent with the subject matter herein, is incorporated in the present disclosure by reference. 7

What is claimed is:

1. In a process for the concentration and decontamination of plutonium present in a dilute solution in the tetravalent state of oxidation together with fission product values, the steps which comprise contacting said solutionwith a bismuth phosphate precipitate, separating said bismuth phosphate precipitate and its associatedplutonium from the supernatant solution, dissolving said separated precipitate and plutonium to form a second solution, oxidizing plutoniumto the hexavalent oxidation state, maintaining said plutonium in the hexavalent state of oxidation while contacting said second solution with a carrier for said fission product values, separating the carrier and associated fission product values from the supernatant solution, reducing the plutonium to the tetravalcnt state, contacting the resulting solution with a carbismuth phosphate precipitate and its associated-pluto- 30 nium from the supernatant solution, dissolving the separated'precipitate and plutonium to form a second solution, oxidizing the plutonium to the heX-avalent oxidation state, maintaining plutonium in said second solution in the hexavalent state while precipitating a bismuth phosphate precipitate, separating said bismuth phosphate precipitate and associated contaminating elements from the supernatant solution, reducing the plutonium to the tetravalent state, precipitating in the resulting solution a carrier for plutonium selected from the group consisting 40 of lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate and niobium oxide, and. separating said carrier and its associated plutonium from the supernatant liquid. '45-. 3... In a process for the concentration and decontami- H lnation ofplutonium present in a dilute solution in the tetravalent state of oxidation together with fission product values, the steps which comprise precipitating in said solution a bismuth phosphate carrier for plutonium, sep- 0 arating the bismuth phosphate carrier and its associated plutonium from the supernatant solution, dissolving the separated'carrier and plutonium to form a second solution, oxidizing the plutonium to the hexavalent oxidation state, maintaining the plutonium in said second solution in the hexavalent state of oxidation while precipitating a second bismuth phosphate carrier, separating said second bismuth phosphate carrier and associated fission products from the supernatant solution reducing the plutonium to the tetravalent state, precipitating in the resulting solution a third carrier selected from the group consistingof lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ccric phosphate, thorium iodate, thorium oxalate, bismuth oxalate, and niobium oxide, separating the third carrier and its associated plutonium from the supernatant liquid, and dissolving the separated third carrier and its associated plutonium to form a third solution having a volume substantially less. than the volume of the initial solution.

, 4. In a process'for the concentration and decontamination of plutonium present in a dilute solution in the tetravalent state together with fission product values, the steps which comprise precipitating in said solution a bismuth phosphate carrier for said plutonium, separating said carrier and associated plutonium from the supcr- 76 natant solution, dissolving the separated carrier and plutonium to form a second solution, oxidizing the plutonium to the hexavalent oxidation state, maintaining the plutonium in the hexavalent state in said second solution while precipitating a second carrier selected from the group consisting of lanthanum fluoride, ceric phosphate, bismuth oxalate, thorium iodate, and thorium oxalate separating the second carrier and associated fission product values from the supernatant solution, reducing said plutonium values to the tetravalent state, precipitating in the resulting solution a third carrier selected from the group consisting of lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate, and niobium oxide, and separating the third carrier and its associated plutonium from the supernatant solution.

- 5. In a process for the concentration and decontamination of plutonium in a dilute solution containing said plutonium in the tetravalent state together with fission product values, the steps which comprise precipitating in said solution a bismuth phosphate carrier for said plutonium, separating said carrier and associated plutonium from the supernatant solution, dissolving the separated carrier and plutonium to form a second solution, oxidizing the plutonium to the hexavalent state of oxidation, maintaining the plutonium in the hexavalent state of oxidation in said second solution while precipitating a second carrier selected from the group consisting of lanthanum fluoride, ceric phosphate, bismuth oxalate, thorium iodate, and thorium oxalate, separating the second carrier and associated fission product values from the supernatant solution, reducing the plutonium values to the tetravalent state, precipitating in the resulting solution a third carrier selected from the group consisting of lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate, and niobium oxide, separating the third carrier and its associated plutonium from the supernatent solution, and dissolving the separated third carrier and its associated plutonium to form a third solution having a volume substantially less than the volume of the initial solution.

6. In a process for the concentration and decontamination of plutonium present in a dilute solution in the tetravalent state together with fission product values, the steps which comprise precipitating in said solution a bismuth phosphate carrier for said plutonium, separating said carrier and its associated plutonium from the supernatant solution, dissolving the separated carrier and plutonium to form a second solution, oxidizing the plutonium to the hexavalent state, maintaining the plutonium in said second solution in the hexavalent state while precipitating a second bismuth phosphate carrier, separating said second carrier and its associated fission product values from the supernatant solution, precipitating in said supernatant solution a third carrier selected from the group consisting of lanthanum fluoride, cen'c phosphate, bismuth oxalate, thorium iodate, and thorium oxalate, separating the third carrier and its associated fission product values from the supernatant solution, reducing the plutonium in said supernatant solution to the tetravalent state, precipitating in the resulting solution a fourth carrier selected from the group consisting of lanthanum fluoride, lanthanum hydroxide, lanthanum phosphate, ceric phosphate, thorium iodate, thorium oxalate, bismuth oxalate, and niobium oxide, and separating the fourth carrier and its associated plutonium from the supernatant solution.

7. In a process for the concentration and decontamination of plutonium present in a dilute solution derived from neutron-irradiated uranium, said solution containing tetravalent plutonium and radioactive uranium fission products, the steps which comprise precipitating bismuth phosphate in said solution, separating the bismuth phosphate and its associated plutonium from the supernatant solution, dissolving the separated bismuth phosphate and plutonium to form a second solution, oxidizing the plutonium in said second solution to the hexavalent state, precipitating bismuth phosphate in the resulting solution, separating the bismuth phosphate and its associated fission products from the supernatant solution, reducing the plutonium in the supernatant solution to the tetravalent state, precipitating lanthanum fluoride in the resulting solution and separating the lanthanum fluoride and its associated plutonium from the supernatant liquid.

8. In a process for the concentration and decontamination of plutonium present in a dilute solution derived from neutron-irradiated uranium, said solution containing tetravalent plutonium and radioactive uranium fission products, the steps which comprise precipitating bismuth phosphate in said solution, separating the bismuth phosphate and its associated plutonium from the supernatant solution, dissolving the separated bismuth phosphate and plutonium to form a second solution, oxidizing the plutonium in said second solution to the hexavalent state, precipitating bismuth phosphate in the resulting solution, separating the bismuth phosphate and its associated fis sion products from the supernatant solution, reducing the plutonium in the supernatant solution to the tetravalent state, precipitating lanthanum fluoride in the re sulting solution, separating the lanthanum fluoride and its associated plutonium from the supernatant liquid, treating the separated fluorides with an alkali metal hydroxide to transform said fluorides to hydroxides, and dissolving the resulting hydroxides to form a solution having a volume substantially less than the volume of the initial solution.

9. In a process for the concentration and decontamination of plutonium present in a dilute solution derived from neutron-irradiated uranium, said solution containing tetravalent plutonium and radioactive uranium fission products, the steps which comprise precipitating bismuth phosphate in said solution, separating the bismuth phosphate precipitate and its associated plutonium from the supernatant solution, dissolving the bismuth phosphate and plutonium to form a second solution, oxidizing the plutonium in said second solution to the hexavalent state, maintaining the plutonium in solution in the hexavalent state while precipitating lanthanum fluoride in said solution, separating the lanthanum fluoride precipitate and its associatedfission products from the supernatant solution, reducing the plutonium'in said supernatant solution to the tetravalent state, precipitating lanthanum fluoride in the resulting solution, and separating the lanthanum fluoride precipitate and its associated plutonium from the supernatant liquid.

10. In a process for the concentration and decontamination of plutonium present in a dilute solution derived from neutron-irradiated uranium, said solution containing tetravalent plutonium and radioactive uranium fission products, the steps which comprise precipitating bismuth phosphate in said solution, separating the bismuth phosphate precipitate and its associated plutonium from the supernatant solution, dissolving the bismuth phosphate and plutonium to form a second solution, oxidizing the plutonium in said second solution to the hexavalent state, maintaining the plutonium in solution in the hexavalent state while precipitating lanthanum fluoride in said solution, Separating the lanthanum fluoride precipitate and its associated fission products from the supernatant solution, reducing the plutonium in said supernatant solution to the tetravalent state, precipitating lanthanum fluoride in the resulting solution, separating the lanthanum fluoride precipitate and its associated plutonium fluoride from the supernatant liquid, treating the separated fluorides with an alkali metal hydroxide to transform the fluorides to hydroxides, and dissolving the resulting hydroxides to form a solution having a volume substantially less than the volume of the initial solution.

. 11. In a process for the concentration and decontamination of plutonium present in a dilute solution derived from neutron-irradiated uranium, said solution containing tetravalent plutonium and radioactive uranium fission products, the steps which comprise precipitating bismuth phosphate in said solution, separating the bismuth phosphate precipitate and its associated plutonium from the supernatant solution, dissolving the bismuth phosphate and plutonium to form a second solution, oxidizing the plutonium in said second solution to the hexavalent state, maintaining the plutonium in solution in the hexavalent. state while precipitating bismuth phosphate in said second solution, separating the bismuth phosphate precipitate and its associated fission products from the supernatant solution, precipitating lanthanum fluoride in said supernatant solution, separating the lanthanum fluoride precipitate and its associated fission products from the supernatant solution, reducing the plutonium in said supernatant solution to the tetravalent state, precipitating lanthanum fluoride in the resulting solution and separating the lanthanum fluoride precipitate and its associated plutonium from the supernatant liquid.

12. In a process for the concentration and decontamination of plutonium present in a dilute solution derived from neutron-irradiated uranium, said solution containing tetravalent plutonium and radioactive uranium fission products, the steps which comprise precipitating bismuth phosphate in said solution, separating the bismuth phosphate precipitate and its associated plutonium from the supernatant solution, dissolving the bismuth phosphate and plutonium to form a second solution, oxidizing the plutonium in said second solution to the hexavalent state, maintaining the plutonium in solution in the hexavalent state while precipitating bismuth phosphate in said second solution, separating the bismuth phosphate precipitate and its associated fission products from the supernatant solution, precipitating lanthanum fluoride in said supernatant solution, separating the lanthanum fluoride precipitate and its associated fission products from the supernatant solution, reducing the plutonium in said supernatant solution to the tetravalent state, precipitating lanthanum fluoride in the resulting solution, separating the lanthanum fluoride precipitate and its associated plutonium fluoride from the supernatant liquid, treating the separated fluorides with an alkali metal hydroxide to transform the fluorides to hydroxides, and dissolving the resulting hydroxides to form a solution having a volume substantially less than the volume of the initial solution.

' 13. In a process for the separation of plutonium from fission product values present in an aqueous solution of an inorganic acid containing plutonium in the hexavalent state, the steps which comprise forming a bismuth phosphate carrier precipitate in said solution, separating said precipitate together with associated fission product values, reducing plutonium in said solution to the tetravalent state, forming a lanthanum fluoride carrier precipitate in said solution, separating said precipitate together with associated plutonium fluoride, metathesizing the mixed fluorides to plutonium and lanthanum hydroxides, forming a solution of said hydroxides, adjusting the acidity of said solution so that plutonium but not lanthanum will precipitate upon the addition of a peroxide, contacting said solution with a peroxide to precipitate the plutonium and separating plutonium peroxide from the solution.

14. In a process for the separation of plutonium from fission product values present in an aqueous solution of an inorganic acid containing plutonium in the hexavalent state, the steps which comprise forming a bismuth phos phate carrier precipitate in said solution, separating said precipitate together with associated fission product values from said solution, reducing plutonium contained in said solution to the tetravalent state, forming a lanthanum fluoride carrier precipitate in said solution, separating said precipitate together with associated plutonium fluoride,

metathesizing the mixed fluorides to plutonium and Ian-.

thanum hydroxides, forming a solution of the hydroxides,

2t) adjusting the pH value of said solution to between approximately l and 2 contacting said solution with a peroxide t p ip t e p ut niu n epa at p utonium peroxide from the solution.

15 The process of recovering plutonium values from a bismuth phosphate precipitate containing said plutoniumrvalues in the tetravalent state, comprising dissolving said precipitate, precipitating a lanthanum fluoride carrier precipitate in the solution formed, separating said precipitate together with associated plutonium fluoride, metathesizing the mixed fluorides to plutonium and lanthanum hydroxides, forming a solution of the hydroxides, adjusting the acidity of said solution so that plutonium but notlanthanurnt will precipitate upon the addition of a peroxide, adding a peroxide to precipitate the plutonium and separating plutonium peroxide from: the solution.

16.7111 a process for the separation of plutonium from. fission product values present in an aqueous solution of an inorganic acid containing plutonium in the hexavalent state, the steps which comprise forming a bismuth phos-. phate carrier precipitate in said solution, separating said precipitate together with associated fission product values, reducing the plutonium contained in said solution to the tetravalent state, forming a lanthanum fluoride carrier precipitate in said solution, separating said precipitate together with associated plutonium fluoride, metathesizing with sodium hydroxide the mixed plutonium and Ianthanum fluorides to the hydroxides, forming a solution of the hydroxides, adjusting the acidity of said solution so that plutonium butrnot lanthanum will precipitate upon the addition of a peroxide, contacting said solution with a peroxide to precipitate the plutonium and separating plutonium peroxide from the solution.

17. In a process for the concentration and decontamination of plutonium present in a dilute solution containing plutonium in the tetravalent state and fission product values, the steps which comprise precipitating in said solution a bismuth phosphate precipitate, separating the bismuth phosphate precipitate and its associated pluto nium from the supernatant solution, dissolving the separated precipitate and plutonium to form a second solution, oxidizing the plutonium to an oxidation state of +6, maintaining plutonium in said second solution in anoxidation state of ,+6 while precipitating a bismuth phosphate precipitate, separating said bismuth phosphate precipitate and associated fission product values from the supernatant solution, reducing the plutonium in said supernatant solution to' an oxidation state of +4, precipitating in the resulting solution a lanthanum phosphate carrier for the plutonium and separating said lanthanum phosphate precipitate and its associated plutonium from the supernatant liquid. 7

' 18. A process of separating tetravalent plutonium values from fission product values contained in an aqueous nitric acid solution, comprising incorporating in said solution a bismuth phosphate precipitate whereby said plutonium values and some fission product values are carried on said bismuth phosphate, while the bulk of said fission product values remains in solution; separating said bismuth phosphate precipitate from said solution; dissolving said precipitate in aqueous nitric acid; oxidizing said plutonium values in the solution formed to the hexavalent state; incorporating ceric phosphate and bismuth phosphate, precipitates in said solution whereby the fission product values present in the solution are carried'on said precipitates, while the plutonium values remain in solution; separating said precipitates from the solution; reducing' the plutonium values in the solution to the tetravalent state; incorporating a thorium iodate precipitate in said solution whereby the plutonium values are carried on said iodate; separating said iodate from the solution; and dissolving the iodate precipitate with aqueous hydrochloric acid.

19. The process of. clairnd8 wherein plutonium is,

21 oxidized to the hexavalent state with a mixture of ceric nitrate and potassium dichromate and the hexavalent plutonium is reduced to the tetravalent state with sodium nitrite.

20. A process of separating tetravalent plutonium val ues from fission product values contained in an aqueous nitric acid solution, comprising incorporating in said solution a bismuth phosphate precipitate whereby said plutonium values and some fission product values are carried on said bismuth phosphate, while the bulk of said fission product values remains in solution; separating said bismuth phosphate precipitate from said solution; dissolving said precipitate in aqueous nitric acid; oxidizing said plutonium values in the solution formed to the hexavalent state; incorporating a bismuth phosphate pre cipitate in said solution whereby fission product values are carried on said precipitate, while the plutonium values remain in solution; separating said bismuth phosphate precipitate from the solution; incorporating a thorium iodate precipitate in said solution whereby remaining fission products are carried on said iodate; separating said thorium iodate precipitate from the plutoniumcontaining solution; reducing the plutonium values in the solution to the tetravalent state; incorporating another thorium iodate precipitate in said solution whereby said plutonium values are carried thereon; removing said plutonium-containing precipitate from the solution; and dissolving the precipitate in aqueous hydrochloric acid.

21. The process of claim 20 wherein tetravalent plutonium is oxidized to the hexavalent state with lead oxide, Pb O and the hexavalent plutonium is reduced to the tetravalent state with sulfur dioxide.

22. A process of separating tetravalent plutonium values from fission product values contained in an aqueous nitric acid solution, comprising incorporating in said solution a bismuth phosphate precipitate whereby said plutonium values and some fission product values are carried on said bismuth phosphate, while the bulk of said fission product values remains in solution; separating said bismuth phosphate precipitate from said solution; metathesizing said precipitate with a mixture of sodium hydroxide and potassium carbonate; dissolving said precipitate in aqueous nitric acid; oxidizing said plutonium values in the solution formed to the hexavalent state; incorporating another bismuth phosphate precipitate in said solution whereby part of the fission products present in the solution are carried thereon, while the plutonium values remain in solution; separating said precipitate from the solution; incorporating a niobium pentoxide precipitate in the solution whereby the remaining fission products are substantially carried on said niobium pentoxide; removing said niobium pentoxide precipitate from the solution; reducing the hexavalent plutonium to the tetravalent state; incorporating another niobium pentoxide precipitate in said solution whereby said plutonium values are carried thereon; separating said niobium pentoxide precipitate from the solution; and dissolving the niobium pentoxide precipitate in aqueous sulfuric acid.

23. The process of claim 22 wherein the tetravalent plutonium is oxidized to the hexavalent state with a mixture of sodium bismuthate and potassium dichromate and the hexavalent plutonium is reduced to the tetravalent state by means of hydrogen peroxide.

24. A process of separating tetravalent plutonium valules from fission product values contained in an aqueous nitric acid solution, comprising incorporating in said solution a bismuth phosphate precipitate whereby said plutonium values and some fission product values are carried on said bismuth phosphate, while the bulk of said fission product values remains in solution; metathesizing said bismuth phosphate precipitate with sodium hydroxide; dissolving said metathesized precipitate in aqueous nitric acid; oxidizing said plutonium values in the solution formed to the hexavalent state; incorporating a mixture of cerous and bismuth phosphates in said solution whereby part of the fission product values present in the solution are carried on said phosphates, while the plutonium values remain in solution; separating said phosphates from the solution; incorporating a niobium pentoxide precipitate in said solution whereby the remaining fission products are substantially carried thereon; removing said niobium pentoxide precipitate from the solution; reducing the hexavalent plutonium in the solution to the tetravalent state; incorporating a lanthanum fluoride precipitate in said solution whereby the plutonium values are carried thereon; metathesizing said lanthanum fluoride precipitate with sodium hydroxide; separating said metathesized precipitate from the solution; and dissolving the precipitate in aqueous nitric acid.

25. The process of claim 24 wherein the tetravalent plutonium is oxidized to the hexavalent state with ceric nitrate and the hexavalent plutonium is reduced to the tetravalent state with hydrogen peroxide.

References Cited in the file of this patent UNITED STATES PATENTS 2,776,185 Werner et a1. Ian. 1, 1957 2,785,951 Thompson et a1 Mar. 19, 1957 2,799,553 Thompson et al. July 16, 1957 2,868,619 Ritter Jan. 13, 1959 2,871,251 Gofman et al. Jan. 27, 1959 OTHER REFERENCES Seaborg et al.: The Transuranium Elements, 1st edition by Seaborg, Katz and Manning. National Nuclear Energy Series, 1949, pages 25-38. Footnote on page 25 states article was prepared on March 19, 1942, and mailed on March 21 to the Uranium Committee in Washington, DC. The former date is relied upon.

Seaborg: Chemical and Engineering News, vol. 23, No. 23, pages 2l90-2193 (1945).

Harvey et al.: Journal of the Chemical Society, August 1947, pages 1010-1021.

Seaborg et al.: Journal of the American Chemical Society, vol. 70, pages 1128-1134 1948). Report submitted March 21, 1942. 

1. IN A PROCESS FOR THE CONCENTRATION AND DECONTAMINATION OF PLUTONIUM PRESENT IN A DILUTE SOLUTION IN THE TETRAVALENT STATE OF OXIDATION TOGETHER WITH FISSION PRODUCT VALUES, THE STEPS WHICH COMPRISE CONTACTING SAID SOLUTION WITH A BISMUTH PHOSPHATE PRECIPITATE, SEPARATING SAID BISMUTH PHOSPHATE PRECIPITATE, SEPARATING TONIUM FROM THE SUPERNATANT SOLUTION, DISSOLVING SAID SEPARATED PRECIPITATE AND PLUTONIUM TO FORM A SECOND SOLUTION, OXIDIZING PLUTONIUM TO THE HEXAVALENT OXIDA TION STATE, MAINTAINING SAID PLUTONIUM IN THE HEXAVALENT STATE OF OXIDATION WHILE CONTACTING SAID SECOND SOLUTION WITH A CARRIER FOR SAID FISSION PRODUCT VALUES, SEPARATING THE CARRIER AND ASSOCIATED FISSION PRODUCT VALUES FROM THE SUPERNATANT SOLUTION, REDUCING THE PLUTONIUM TO THE TETRAVALENT STATE, CONTACTING THE RESULTING SOLUTION WITH A CARRIER FOR SAID PLUTONIUM SELECTED FROM THE GROUP CONSISTING OF LATHANUM FLUORIDE, LANTHANUM HYDROXIDE, LANTHANUM PHOSPHATE, CERIC PHOSPHATE, THORIUM IODATE, THORIUM OXALATE, BISMUTH OXALATE, AND NIOBIUM OXIDE, SEPARATING THE CARRIER AND ASSOCIATED PLUTONIUM FROM THE SUPERNATANT SOLUTION, AND DISSOLVING THE SEPARATED CARRIER AND PLUTONIUM TO FORM A THIRD SOLUTION HAVING A VOLUME SUBSTANTIALLY LESS THAN THE VOLUME OF THE INITIAL SOLUTION. 